Radiation damage in helium ion-irradiated reduced activation ferritic/martensitic steel

doi:10.1016/j.net.2017.10.012

	Radiation damage in helium ion-irradiated reduced activation ferritic/martensitic steel
	Xia, L. D.1; Liu, W. B.1,2; Liu, H. P.3; Zhang, J. H.2; Chen, H.1; Yang, Z. G.1; Zhang, C.1
刊名	NUCLEAR ENGINEERING AND TECHNOLOGY
	2018-02-01
卷号	50 页码:132-139
关键词	Grain Boundary Helium Bubble Irradiation Temperature Reduced Activation Ferritic/Martensitic Steel
ISSN号	1738-5733
DOI	10.1016/j.net.2017.10.012
英文摘要	Nanocrystalline reduced activation ferritic/martensitic (RAFM) steel samples were prepared using surface mechanical attrition treatment (SMAT). Un-SMATed and SMATed reduced activation ferritic/martensitic samples were irradiated by helium ions at 200 degrees C and 350 degrees C with 2 dpa and 8 dpa, respectively, to investigate the effects of grain boundaries (GBs) and temperature on the formation of He bubbles during irradiation. Experimental results show that He bubbles are preferentially trapped at GBs in all the irradiated samples. Bubble denuded zones are clearly observed near the GBs at 350 degrees C, whereas the bubble denuded zones are not obvious in the samples irradiated at 200 degrees C. The average bubble size increases and the bubble density decreases with an increasing irradiation temperature from 200 degrees C to 350 degrees C. Both the average size and density of the bubbles increase with an increasing irradiation dose from 2 dpa to 8 dpa. Bubbles with smaller size and lower density were observed in the SMATed samples but not in the un-SMATed samples irradiated in the same conditions, which indicate that GBs play an important role during irradiation, and sink strength increases as grain size decreases. (c) 2017 Korean Nuclear Society, Published by Elsevier Korea LLC.
资助项目	National Magnetic Confinement Fusion Energy Research Project[2015GB118001] ; China Postdoctoral Science Foundation[2015M582669] ; Fundamental Research Funds for the Central Universities ; [2016YFB0300104]
WOS关键词	AUSTENITIC STAINLESS-STEEL ; MARTENSITIC STEEL ; VOID FORMATION ; GRAIN-SIZE ; BCC FE ; NEUTRON ; MICROSTRUCTURE ; 1ST-PRINCIPLES ; MECHANISM ; EXPOSURE
WOS研究方向	Nuclear Science & Technology
语种	英语
出版者	KOREAN NUCLEAR SOC
WOS记录号	WOS:000425255700014
资助机构	National Magnetic Confinement Fusion Energy Research Project ; China Postdoctoral Science Foundation ; Fundamental Research Funds for the Central Universities
内容类型	期刊论文
源URL	[http://119.78.100.186/handle/113462/46306]
专题	中国科学院近代物理研究所
通讯作者	Liu, W. B.; Zhang, C.
作者单位	1.Tsinghua Univ, Minist Educ, Sch Mat Sci & Engn, Key Lab Adv Mat, Beijing 100084, Peoples R China 2.Xi An Jiao Tong Univ, Dept Nucl Sci & Technol, Xian 710049, Shaanxi, Peoples R China 3.Chinese Acad Sci, Inst Modern Phys, Lanzhou 730000, Gansu, Peoples R China
推荐引用方式 GB/T 7714	Xia, L. D.,Liu, W. B.,Liu, H. P.,et al. Radiation damage in helium ion-irradiated reduced activation ferritic/martensitic steel[J]. NUCLEAR ENGINEERING AND TECHNOLOGY,2018,50:132-139.
APA	Xia, L. D..,Liu, W. B..,Liu, H. P..,Zhang, J. H..,Chen, H..,...&Zhang, C..(2018).Radiation damage in helium ion-irradiated reduced activation ferritic/martensitic steel.NUCLEAR ENGINEERING AND TECHNOLOGY,50,132-139.
MLA	Xia, L. D.,et al."Radiation damage in helium ion-irradiated reduced activation ferritic/martensitic steel".NUCLEAR ENGINEERING AND TECHNOLOGY 50(2018):132-139.